[Federal Register Volume 63, Number 135 (Wednesday, July 15, 1998)]
[Notices]
[Pages 38198-38214]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-18684]



[[Page 38198]]

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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 
1954, as amended (the Act), to require the Commission to publish notice 
of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 22, 1998, through July 2, 1998. The 
last biweekly notice was published on July 1, 1998 (63 FR 35986).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By August 14, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.

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    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: May 29, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to credit the automatic function of 
the pressurizer power operated relief valves (PORVs) to provide 
mitigation for inadvertent safety injection at power accident. The 
limiting condition for operation and surveillance requirements for the 
PORVs would also be revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The changes to the Technical Specification (TS) Limiting Condition 
for Operation (LCO), Surveillance Requirements, and Bases do not 
involve an increase in the probability or consequences of the 
Inadvertent Operation of Emergency Core Cooling System (Spurious SI) at 
Power transient. Crediting the PORVs in the maximum pressurizer 
overfill case for this transient does not increase the probability of 
the occurrence of the transient since the automatic function of the 
PORVs for Reactor Coolant System (RCS) pressure control is not an 
initiator for the Spurious SI at Power transient. This change allows 
for the NRC Standard Review Plan (NUREG-0800) acceptance criteria to be 
met for the Spurious SI at Power transient, ensuring that the 
consequences of this transient remain within acceptable levels.
    As documented in various Safety Evaluation Reports (SERs) from the 
NRC, the overpressure protection function of the PORVs was not 
originally considered to be a safety related function. In response to 
Generic Issue 70, the NRC performed a regulatory analysis related to 
PORV and block valve reliability in Pressurized Water Reactor (PWR) 
plants. This regulatory analysis is documented in NUREG-1316, 
``Technical Findings and Regulatory Analysis Related to Generic Issue 
70, Evaluation of Power-Operated Relief Valve and Block Valve 
Reliability in PWR Nuclear Power Plants,'' where the NRC staff 
concluded that it was not cost effective to backfit non-safety related 
PORVs to upgrade them to safety related status to perform safety 
related functions. The safety related functions were those detailed in 
Section 2.1 of NUREG-1316 and any other safety related function 
identified in the future. As an example, the PORVs are credited for the 
cold overpressure protection function of the reactor pressure vessel 
during low temperature operations. The analysis documented in this 
License Amendment request demonstrates that the PORVs provide an 
acceptable level of quality and performance to allow them to be 
credited to mitigate the consequences of the Spurious SI at Power 
transient documented in Byron and Braidwood Updated Final Safety 
Analysis Report (UFSAR) Section 15.5.1. The PORVs are equipped with 
safety related actuators and safety related accumulator tanks which 
maintain valve function during a loss of instrument air. The position 
indication and control switches in the Main Control Room (MCR) are 
safety related. All pressurizer PORV open/close functions and circuitry 
are supplied with uninterruptible Class 1E power supplies. The 
automatic portion of the PORV circuitry which processes the high 
pressurizer and high RCS pressure at low temperature is designated non-
safety related and is isolated from the safety related portions of the 
circuitry by safety related interposing relays which actuate on a 
faulted condition. However, both Byron and Braidwood Stations have 
implemented modifications for both Units 1 and 2, which ensure that 
automatic control of both PORVs is available during loss of offsite 
power conditions. In addition, the PORV function is monitored within 
the scope of the Maintenance Rule Program and the postulated failure of 
the PORV automatic function does not result in unacceptable risk.
    The probability of a Spurious SI at Power transient is not affected 
by this proposed change and the above analysis demonstrates that the 
PORVs will adequately function in automatic mode to mitigate the 
consequences of the transient. As such, there are no changes in the 
type or amount of any effluent released offsite as a result of this 
change. Therefore, based on this evaluation, this proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

[[Page 38200]]

    This proposed change does not create the possibility of a new or 
different accident from any accident previously evaluated. This change 
would specifically allow for the PORV automatic function to be credited 
in Modes 1, 2, and 3 for the Spurious SI at Power transient only. This 
change allows for added assurance that the acceptance criteria as 
documented in the NRC Standard Review Plan (NUREG-0800) for ANS 
Condition II transients will be met. The acceptance criteria of concern 
is that a Condition II transient must not lead to an event (Condition 
III or IV) of more significant consequences without additional failures 
occurring. The PORV automatic function is to be credited with 
mitigating the maximum pressurizer overfill case for the Spurious SI at 
Power transient. This case has the acceptance criteria that the 
pressurizer must not go water solid prior to RCS pressure reaching the 
setpoint of the pressurizer safety relief valves (PSRVs). This 
conservative acceptance criteria is based on the fact that the PSRVs 
are not qualified to pass subcooled water and reseat, thereby creating 
a concern for an uncontrolled release path from the RCS. This proposed 
change helps ensure that the acceptance criteria for this accident are 
met. There is a small probability that the PORV function, either 
automatic or manual, would not successfully mitigate this transient due 
to the failure of one or both PORVs. However, the low likelihood of a 
total failure of the PORV function during the Spurious SI at Power 
transient does not create a new accident because a similar scenario is 
already addressed by UFSAR Section 15.6.1, ``Inadvertent Opening of a 
Pressurizer Safety or Relief Valve.'' The UFSAR analysis for the 
Section 15.6.1 ANS Condition II transient indicates that the 
radiological consequences of this transient are significantly less than 
that of a LOCA and are therefore, acceptable. The same arguments for 
radiological consequences apply to the Spurious SI at Power transient 
in the event the PORV automatic function fails and water relief occurs 
through the PSRVs.
    The proposed change to the LCO requirements in TS Section 3/4.4.4 
would allow for the PORV block valve to be closed but remain energized 
in the event a PORV was considered inoperable due to the automatic 
actuation circuitry. Currently, the PORV block valve is closed but 
remains energized only if a PORV is considered inoperable due to 
excessive seat leakage. The proposed change would extend the allowance 
to include the circumstance where the PORV was inoperable due to the 
automatic actuation circuitry. This allows a PORV to remain functional 
in the manual mode for other safety related functions consistent with 
the discussion contained in NRC NUREG-1316. However, this revised LCO 
requirement would not represent a new failure mode or accident over 
what has been previously evaluated.
    In summary, the proposed changes documented in this TS amendment to 
credit the automatic PORV function and to revise the TS LCO 
requirements for PORV inoperability do not create the potential for any 
new or different accidents from what was previously evaluated.
    3. The change does not involve a significant reduction in a margin 
of safety.
    The current TS bases do not credit the function of the pressurizer 
PORVs for any Mode 1, 2, or 3 transients. This change would allow for 
the PORV automatic function to be credited for the Spurious SI at Power 
transient only. This does not represent a significant reduction in the 
margin of safety. This change would allow for the conservative 
acceptance criteria for the current UFSAR design analysis to be met. 
The PORVs are reliable and are maintained in a manner consistent with 
their proposed safety related function to mitigate the Spurious SI at 
Power transient. This proposed change would not result in a significant 
increase in risk or consequences, and therefore, does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: May 28,1998.
    Description of amendment request: The proposed amendment proposes 
changes to the Final Safety Analysis Report (FSAR) to include a 
description of the use of Generic Letter (GL) 87-11, ``Relaxation in 
Arbitrary Intermediate Pipe Rupture Requirements,'' and NUREG/CR-2913, 
``Two-Phase Jet Loads,'' as a part of the approved licensing basis and 
design basis for Crystal River Unit 3. GL 87-11 will be used to 
determine where high energy line breaks (HELB) are postulated to occur 
for high energy lines located inside the Reactor Building (RB) and 
analyzed in accordance with the guidelines described in USAS B31.1.0-
1967, ``USA Standard Code for Pressure Piping, Power Piping.'' NUREG/
CR-2913 will be used to determine the effects of the resultant jet 
impingement from postulated Reactor Coolant System (RCS) piping 
ruptures on safety-related systems, structures, and components (SSCs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The use of new design methodologies for determining postulated 
break locations of RCS piping and other high energy lines located 
inside containment, and the dynamic effects of postulated ruptures of 
RCS piping on SSCs required for safe shutdown or accident mitigation, 
does not impact the design of these high energy lines such that 
previously analyzed ruptures would now be more likely to occur. The 
approval of the license amendment will not result in an actual 
modification to RCS piping or other high energy lines which would 
reduce their design capabilities to maintain pressure boundary 
integrity during normal operating and accident conditions. By using 
these new design methodologies, protection of SSCs required for 
accident mitigation is assured. Protection of SSCs required for 
accident mitigation will continue to be assured by use of these well-
defined design methodologies if modifications to those SSCs are 
implemented in the future. Therefore, there will be no reduction in the 
capability of those SSCs in limiting the consequences of previously 
evaluated accidents, and the proposed amendment does not significantly 
increase the probability or consequence of an accident previously 
evaluated.

[[Page 38201]]

    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    The use of new design methodologies for determining postulated 
break locations of RCS piping and other high energy lines located 
inside containment, and the dynamic effects of postulated ruptures of 
RCS piping on SSCs required for safe shutdown or accident mitigation, 
does not impact the design of these high energy lines such that 
previously unanalyzed ruptures would now occur. The approval of the 
license amendment will not result in an actual modification to RCS 
piping or other high energy lines which would reduce their design 
capabilities to maintain pressure boundary integrity during normal 
operating and accident conditions. By using these new design 
methodologies, the current design of RCS piping and other high energy 
lines located inside containment can be shown to include sufficient 
design margin to prevent unanalyzed ruptures from occurring. Therefore, 
use of these design methodologies instead of the previous licensing 
basis requirements cannot create the possibility of a new or different 
kind of accident.
    3. Involve a significant reduction in a margin of safety?
    The use of new design methodologies for determining postulated 
break locations of RCS piping and other high energy lines located 
inside containment, and the dynamic effects of postulated ruptures of 
RCS piping on SSCs required for safe shutdown or accident mitigation, 
does not impact the design of these high energy lines such that 
unanalyzed ruptures would now occur, and cannot create a reduction in 
the margin of safety for those ruptures of high energy lines previously 
analyzed. The approval of the license amendment will not result in an 
actual modification to RCS piping or other high energy lines which 
would reduce their design capabilities to maintain pressure boundary 
integrity during normal operating and accident conditions. By using 
these new design methodologies, protection of SSCs required for 
accident mitigation is assured. Protection of SSCs required for 
accident mitigation will continue to be assured by use of these well-
defined design methodologies if modifications to those SSCs are 
implemented in the future. Therefore, the capability of those SSCs to 
limit the consequences of previously evaluated accidents at levels 
below the approved acceptance limits will continue to be assured. As a 
result, use of these design methodologies instead of the previous 
licensing basis and design basis requirements cannot significantly 
reduce the existing margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: May 28, 1998.
    Description of amendment request: Revision of Technical 
Specification (TS) 4.5.A.1 such that the first Type A test required by 
the primary containment leakage rate testing program be performed 
during refueling outage 18 rather than refueling outage 17.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed changes does not alter the design, function or manner 
of operation of any structures, systems or components. As a result, the 
proposed change does not affect any of the parameters or conditions 
that could contribute to initiation of any accidents.
    NUREG-1493 found that the effect of containment leakage on overall 
accident risk is small since risk is dominated by accident sequences 
that result in failure or bypass of the containment. The major 
contributor to the total identified leakage from Primary Containment 
comes from Type B and C tested components. Only a small portion of the 
total leakage is detectable soley through Type A testing. The leaks 
that have been found by Type A tests have been only marginally above 
existing requirements. In addition, Oyster Creek has two means 
(monitoring nitrogen use and performing torus to drywell vacuum breaker 
leak tests) of detecting gross containment leakage. The proposed change 
does not alter the requirements to perform Type B and C testing in 
accordance with the Primary Containment Leakage Rate Testing Program 
and does not affect the ability of the facility to mitigate the 
consequences of an accident.
    Therefore, the proposed TS change does not involve an significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Deferring the Type A test for an operating cycle does not alter the 
design, function or manner of operation of any structures, systems or 
components. The proposed change does not affect any of the parameters 
or conditions that could contribute to initiation of any accidents nor 
does it introduce any new mechanisms which could contribute to the 
creation of a new or different kind of accident than previously 
evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed change does not alter the design, function or manner 
of operation of any structures, systems or components. The proposed 
change does not impact the primary containment system's ability to 
provide a barrier against the uncontrolled release of fission products 
in the event of a break in the reactor coolant system nor does the 
proposed change impact the primary containment accident leak rate. In 
addition, NUREG-1493's Summary of Technical Findings states ``Reducing 
the frequency of Type A tests (ILRTs) from the current three per 10 
years to one per 20 years was found to lead to an imperceptible 
increase in risk. The estimated increase in risk is very small because 
ILRTs identify only a few potential containment leakage paths that 
cannot be identified by Type B and C testing, and the leaks that have 
been found by Type A tests have been only marginally above existing 
requirements.'' Therefore, the proposed TS change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library,

[[Page 38202]]

Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, Pitman, 
Poets & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: June 10, 1998.
    Description of amendment request: The proposed revision to the 
Millstone Unit 3 licensing basis would address post-accident mitigation 
activities, vital area access travel routes, and time. NNECO determined 
that the Final Safety Analysis Report (FSAR) description of post-
accident vital area routing was out of date because the radiological 
control area boundary fence created an access problem on the designated 
routes to the hydrogen recombiner and fuel building. The FSAR change 
would revise the routes to accommodate the fence location and allow for 
the time to unlock gates.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 10 CFR 
50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not satisfied. The 
proposed revision does not involve an SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    Final Safety Analysis Report (FSAR) Section 12.3.1.3.2, Post-
accident access to vital areas, and its associated Figures and Tables 
are being updated. The current FSAR descriptions are out of date and as 
such do not include all required post-accident actions. Therefore, this 
FSAR change adds actions to those listed in the FSAR as well as 
incorporating the recalculation of the doses associated with the 
required post-accident actions. The dose calculations utilize the 
appropriate post-accident source terms, area access requirements and 
stay times, including the appropriate routes to the areas. The 
calculations show that for all design basis required post-accident 
actions the calculated dose to the Operators/Emergency workers 
performing those actions remains below the 5 rem criterion of General 
Design Criteria (GDC) 19. The revision to the FSAR provides the 
required post-accident required operator actions. Changing the FSAR to 
include the current post-accident vital access requirements and 
associated information for the supporting dose calculations [cannot] 
cause an accident. In addition, the calculated dose to the Operators/
Emergency workers for all design basis required actions is below the 
GDC 19 limit of 5 rem.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change is to the calculated post-accident vital access dose 
analyses and the FSAR description of that analyses. No new procedural 
Operator/Emergency worker actions are associated with the change.
    However, since the information in the FSAR was outdated, there are 
Operator/Emergency actions being added to the FSAR. Dose calculations 
associated with those actions have been performed utilizing the 
appropriate assumptions with respect to source terms, vital area access 
travel routes and stay times, and times when the post-accident actions 
would be performed. The analyses confirmed that the calculated doses 
associated with all required post-accident actions are less than the 5 
rem limit of GDC 19. There are no changes to the Emergency Operating 
Procedures associated with this change.
    Therefore, the proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The dose calculations confirm that the calculated dose associated 
with all design basis post-accident Operator/Emergency worker actions 
is below the limit of 5 rem of GDC 19. There is one action, initiation 
of hydrogen purge, for which the calculated dose to the Operator/
Emergency worker exceeds 5 rem. This action is a backup means of 
limiting the hydrogen concentration inside containment post-accident. 
This action would only be performed for multiple failures which would 
disable both trains of the safety-grade hydrogen recombiner system. As 
such this action is not a required design basis action and does not 
need to meet the 5 rem limit. The calculated dose for this action is 
below the 25 rem limit that is specified in the Station Emergency Plan 
for severe accident mitigation actions.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

PECO Energy Company, Public Service Electric and Gas Company, Demarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: February 4, 1998.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) Surveillance Requirement (SR) 
concerning Secondary Containment doors at Peach Bottom Atomic Power 
Station, Units 2 and 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    TS SR 3.6.4.1.2 will be revised to require either all inner or 
outer secondary containment access doors to be closed in each air lock. 
This revision will not adversely affect the ability of the Secondary 
Containment to mitigate the radiological consequences of a Loss-of-
Coolant Accident or fuel handling accident, and does not involve a

[[Page 38203]]

significant increase in the probability or consequences of an accident 
previously evaluated. During those times that one or more inner (or 
outer) doors are open, the closed outer (or inner) doors will serve as 
the Secondary Containment boundary.
    Allowing certain inner or outer Secondary Containment access doors 
in an air lock to be open does not compromise the design of the 
Secondary Containment. No commitment is made in the UFSAR to consider 
the single failure of passive structural components such as Secondary 
Containment doors. As discussed in Section 1.5 of the UFSAR, ``* * * 
Essential safety actions shall be carried out by equipment of 
sufficient redundance and independence that no single failure of active 
components can prevent the required actions''. The same UFSAR section 
goes on to state that, ``For systems or components to which IEEE-279 
(1968) is applicable, single failures of passive electrical components 
are considered, as well as single failures of active components, in 
recognition of the higher anticipated failure rates of passive 
electrical components relative to passive mechanical components.'' 
Therefore, based on this UFSAR discussion, it is concluded that failure 
of outer (inner) secondary containment doors need not be postulated 
with the inner (outer) door being open.
    The performance of the Secondary Containment and the Standby Gas 
Treatment System is unaffected by this activity. Surveillance testing 
will prove the capability to maintain Secondary Containment with only 
inner or only outer doors closed. This change will not result in 
greater or more frequent loading of Secondary Containment doors, and 
does not result in changes that impact the reliability of the Secondary 
Containment and the Standby Gas Treatment System.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The Secondary Containment, in conjunction with the Standby Gas 
Treatment System, provides the means for mitigating the radiological 
consequences of an accident. The configuration of the Secondary 
Containment has no effect on accident initiators which lead to a new or 
different kind of accident. This change will not involve any changes to 
plant systems, structures, or components which could act as new 
accident initiators. The design, function, and reliability of Secondary 
Containment and the Standby Gas Treatment System are also not impacted 
by this change. Therefore, this change will not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    No margins of safety are reduced as a result of this change to the 
TS. No safety limits will be changed as a result of this TS change. The 
Secondary Containment will continue to perform its intended safety 
function of limiting the ground level release of airborne radioactive 
materials and to provide a means for controlled elevated release of the 
building atmosphere so that off-site doses from the postulated design 
basis accidents are below the limits of 10 CFR 100. The design and 
reliability of the Secondary Containment are also not impacted as a 
result of this change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Robert A. Capra.

Power Authority of the State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: November 13, 1997.
    Description of amendment request: The proposed amendment will 
reduce the maximum test interval from 1 year to 6 months for the test 
frequency of the main turbine stop and control valves (TS & CVs) in 
Table 4.1-3 and add a footnote.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
Response
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed change increases the frequency of testing of 
the TS & CVs by reducing the maximum allowable test interval. The 
maximum test interval is reduced from one year to six months. Thus, the 
proposed change will make the maximum test interval more conservative. 
Therefore, the proposed change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
Response
    The proposed license amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed change does not involve the addition of any new 
or different type of equipment, nor does it involve the operation of 
equipment required for safe operation of the facility in a manner 
different from those addressed in the Final Safety Analysis Report.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
Response
    The proposed license amendment does not involve a significant 
reduction in a margin of safety. The proposed change does not adversely 
affect performance of any safety related system or component, 
instrument operation, or safety system setpoints and does not result in 
increased severity of any accidents considered in the safety analysis. 
The proposed change does not reduce the frequency of testing of these 
valves but updates the methodology for determination of the test 
frequency and reduces the maximum test interval from one year to six 
months. It establishes a more conservative acceptance criteria of 5.0 
x  10<SUP>-6</SUP> per year than the NRC acceptance criteria of 1.0  x  
10<SUP>-5</SUP> for a turbine missile event. Therefore, the proposed 
change does not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 38204]]

    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: June 16, 1998.
    Description of amendment request: The proposed amendment would 
relocate the Safety Review Committee review, audit and related record 
keeping requirements from the Technical Specifications (TSs) to Chapter 
17 of the Final Safety Analysis Report (FSAR) (i.e., Quality Assurance 
Program).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
analyzed?
Response
    This amendment application does not involve a significant increase 
in the probability or consequences of an accident previously analyzed. 
The relocation of the SRC [Safety Review Committee] review, audit, and 
related record keeping requirements from the TS to the FSAR does not 
alter the performance or frequency of these activities. Future changes 
to the QA [Qualify Assurance] program, located in Chapter 17 of the 
FSAR, which constitute a reduction in commitments, are governed by 10 
CFR 50.54(a). Therefore, sufficient controls for these requirements 
exist and these changes do not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
Response
    This amendment application does not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The proposed changes involve the relocation of SRC requirements from 
the TS to the FSAR. Relocation of these requirements does not affect 
plant equipment or the way the plant operates. The reviews, audits, and 
record keeping will continue to be performed in the identical manner as 
they are currently being performed. Therefore, the proposed revisions 
cannot create a new or different kind of accident.
    (3) Does the proposed amendment involve a significant reduction in 
a margin of safety?
Response
    This amendment application does not involve a significant reduction 
in a margin of safety. The requested Technical Specification revisions 
relocate SRC review, audit and related record keeping requirements from 
the TS to the FSAR. These requirements are not being altered by this 
relocation. The reviews, audits, and record keeping will continue to be 
performed in the identical manner as they are currently being 
performed. Any changes to these requirements which constitute a 
reduction in commitments will be processed in accordance with 10 CFR 
50.54(a). Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 30, 1998 (TS 98-01).
    Brief description of amendments: The amendments would change the 
Sequoyah (SQN) Technical Specifications (TSs) to allow surveillance 
testing of the reactor coolant system (RCS) pressurizer power-operated 
relief valves (PORVs) in Modes 3, 4, and 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TSs, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The possibility of occurrence or the consequences for an accident 
or malfunction of equipment is not increased as the test conditions for 
the PORVs in Mode 5 are representative conditions based on a steam 
bubble being present, and testing in this mode is more conservative, if 
RCS pressure is less, since there is less fluid force to aid the 
solenoid force in opening the valve. Testing in Modes 3 and 4 was the 
initial request of GL [Generic Letter] 90-06. No changes are proposed 
to operation of the PORV block valves. Offsite dose consequences are 
unchanged by this request.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    A possibility for an accident or malfunction of a different type 
than any evaluated previously in SQN's Final Safety Analysis Report is 
not created; nor is the possibility for an accident or malfunction of a 
different type. A new test method is not required. No new failure modes 
are introduced.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The margin of safety has not been reduced for testing in Mode 5 
since the proposed test conditions are equal to or more conservative, 
if RCS pressure is less, than those currently in use with existing SRs 
[surveillance requirements]. Testing in Modes 3 and 4 was the initial 
request of GL 90-06. The results of the accident analysis remain 
unchanged by this request.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

[[Page 38205]]

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 26, 1998 (TS 98-02).
    Brief description of amendments: The amendments would change the 
Sequoyah Nuclear Plant (SQN) Technical Specifications (TS) and their 
Bases to lower the specific activity of the primary coolant from 1.0 
microcurie/gram dose equivalent iodine-131 to 0.35 microcurie/gram, as 
provided for in NRC Generic Letter 95-05, ``Voltage-Based Repair 
Criteria for Westinghouse Steam Generator Tubes Affected by Outside 
Diameter Stress Corrosion Cracking.'' This change allows a proportional 
increase in main steam line break induced primary-to-secondary leakage 
when implementing the alternate steam generator tube repair criteria, 
which the NRC has already approved for Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TS [or operating 
license(s)], does not involve a significant hazards consideration. 
TVA's conclusion is based on its evaluation, in accordance with 10 CFR 
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed TS change lowers the [maximum allowable] reactor 
coolant specific activity, which allows an increase in the leakage 
quantity that would be postulated to occur during a MSLB accident. This 
in turn allows a larger quantity of tubes with axial ODSCC to remain in 
service. The methodology for identifying and defining the ODSCC and for 
developing the leakage quantity remains unchanged. Therefore, the 
proposed change does not result in a significant increase in the 
probability of an accident.
    An increase in the consequences of an accident would not occur 
because the proportional decrease in reactor coolant specific activity, 
while proportionally increasing the primary-to-secondary leakage during 
a postulated MSLB accident, has been evaluated to confirm the amount of 
activity released to the environment remains unchanged. The evaluation 
uses the same methodology used to establish the original primary-to-
secondary leak limits in [Westinghouse Topical Report] WCAP-13990.
    The control room dose, the low population zone dose, and the dose 
at the exclusion area boundary remains bounded by the acceptance 
criteria of NUREG-0800 and continue to satisfy an appropriate fraction 
of the 10 CFR 100 dose limits and GDC [General Design Criterion] 19. 
Therefore, the proposed TS change does not result in a significant 
increase in the consequences of an accident previously analyzed.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed TS change does not alter the configuration of the 
plant. The changes do not directly affect plant operation. The change 
will not result in the installation of any new equipment or systems or 
the modification of any existing equipment or systems. No new operating 
procedures, conditions or modes will be created by this proposed 
change. SG [steam generator] tube structural integrity, as defined in 
draft Regulatory Guide 1.121, remains unchanged.
    Therefore the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    Lowering the reactor coolant specific activity, while allowing the 
proportional increase in the primary-to-secondary leakage during a 
postulated MSLB accident, keeps the amount--of activity released to the 
environment unchanged. Design basis and offsite dose calculation 
assumptions remain satisfied. Therefore, the proposed change does not 
result in a significant reduction in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: February 27, 1998 (TXX-98033), June 10, 
1998 (TXX 98145).
    Brief description of amendments: The proposed amendment would 
increase the RWST Low-Low level setpoint from ``greater than or equal 
to 40%'' to ``greater than or equal to 45%'' of span for CPSES, Units 1 
and 2. The change raises the RWST Low-Low level setpoint in order to 
increase the volume available to complete containment spray switchover 
without turning off the containment spray pumps.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The changes in the License Amendment Request proposes more 
restrictive setpoint Allowable Values for the RWST Low-Low setpoint. 
This more restrictive value assures that all applicable safety analysis 
limits are being met. Changing an RWST Low-Low setpoint from greater 
than or equal to 40% to greater than or equal to 45% in the Technical 
Specifications has no impact on the probability of occurrence of any 
accident previously evaluated. None of the accident analyses were 
affected, therefore, the consequences of all previously evaluated 
accidents remain unchanged.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the use of a more conservative value 
for the RWST Low-Low setpoint. As such, none of the changes effect 
plant hardware or the operation of plant systems in a way that could 
initiate an accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    There were no changes made to any of the accident analyses or 
safety analysis limits as a result of this proposed change. Further, 
the proposed

[[Page 38206]]

change does not affect the acceptance criteria for any analyzed event. 
ECCS, Containment spray, and the RWST will remain capable of performing 
their safety function, and the new requirement will continue to provide 
adequate assurance of that capability. Raising the RWST Low-Low 
setpoint from 40% to 45% has no impact on the assumptions used in the 
safety analysis as discussed in Chapter 15 of the FSAR. The margin of 
safety established by the Limiting Conditions for Operation also 
remains unchanged. Thus there is no effect on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036.
    NRC Project Director: John N. Hannon.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: June 19, 1998. This amendment request 
supersedes the November 5, 1997, submittal in its entirety (63 FR 
19981).
    Description of amendment request: The proposed Operating License 
change and changes to the technical specifications (TS) would permit 
the use of a temporary alternate supply line (jumper) to provide 
service water (SW) to the component cooling heat exchangers. The 
temporary jumper will permit maintenance to be performed on the 
existing supply line.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Virginia Electric and Power Company has reviewed the proposed 
changes against the criteria of 10 CFR 50.92 and has concluded that the 
changes do not pose a significant safety hazards consideration as 
defined therein. The proposed Operating License and Technical 
Specifications and Bases changes are necessary to allow the use of a 
temporary, seismic, non-missile protected jumper to provide service 
water (SW) to the Component Cooling Heat Exchangers (CCHXs) while 
maintenance work is performed on the existing SW supply line to the 
CCHXs. Since there is only one SW supply line to the CCHXs, an 
alternate SW supply must be provided whenever the line is removed from 
service. The temporary jumper provides this function. The jumper will 
only be used for a 35-day period during each of two Unit 1 refueling 
outages.
    The use of the temporary jumper has been thoroughly evaluated, and 
appropriate constraints and compensatory measures (including a 
Contingency Action Plan) have been developed to ensure that the 
temporary jumper is reliable, safe, and suitable for its intended 
purpose. A complete and immediate loss of SW supply to the operating 
CCHXs is not considered credible, given the project constraints and the 
unlikely probability of a generated missile or heavy load drop. 
Existing station abnormal procedures already address a loss of 
component cooling, and the use of alternate cooling for a loss of decay 
heat removal, in the unlikely event that they are required. 
Furthermore, appropriate mitigative measures have been identified to 
address potential flooding concerns. The minor administrative changes 
merely correct a table format inconsistency and update Basis section 
references.
    Consequently, the operation of Surry Power Station with the 
proposed amendment and license condition will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The SW and CC Systems will function as designed under the Unit 
operating constraints specified by this project (i.e., Unit 2 in 
operation and Unit 1 in a refueling outage), and the potential for a 
loss of component cooling is already addressed by Station Abnormal 
Procedures. Therefore, there is no increase in the probability of an 
accident previously evaluated. The possibility of flooding due to 
failure of the temporary SW supply jumper in the Turbine Building 
basement has been evaluated and dispositioned by the implementation of 
appropriate precautions and compensatory measures to preclude damage to 
the temporary jumper and to respond to a postulated flooding event. A 
flood watch will be present around-the-clock with authority and 
procedural guidance to isolate the jumper, if required. Furthermore, 
the CCHXs serve no design basis accident mitigating function. 
Therefore, the consequences of an accident previously evaluated are not 
increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The SW and CC Systems' design functions and basic configurations 
are not being altered as a result of using a temporary SW supply 
jumper. The temporary jumper is designed to be safety-related and 
seismic with all of the design attributes of the normal SW supply line, 
except for the automatic isolation function and complete missile and 
heavy load drop protection. The design functions of the SW and CC 
systems are unchanged as a result of the proposed changes due to (1) 
required plant conditions, (2) compensatory measures, (3) a Contingency 
Action Plan for restoration of the normal SW supply if required, and 
(4) strict administrative control of the temporary SW isolation valve 
to preclude flooding or to isolate non-essential SW within the design 
basis assumed time limits. Unit 1 will be in a plant condition which 
will provide adequate time to restore the normal SW supply, if 
required. Therefore, since the SW and CC systems will basically 
function as designed and will be operated in their basic configuration, 
the possibility of a new or different type of accident than previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report] is not 
created.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety as defined in the Technical Specifications is 
not reduced since an operable SW flowpath to the required number of 
CCHXs is provided, and Unit operating constraints, compensatory 
measures and contingencies will be implemented as required to ensure 
the integrity and the capability of the SW flowpath. The use of the 
temporary jumper will be limited to the time period when missile 
producing weather is not expected, and Unit 1 meets specified unit 
conditions. Therefore, the temporary SW jumper, under the imposed 
project constraints and compensatory measures, provides the same 
reliability as the normal SW supply line. Furthermore, the 
Probabilistic Safety Assessment for Surry Power Station has been 
reviewed relative to the use of the temporary SW jumper. It has been 
determined that due to the SW restoration project's compensatory and 
contingency measures, as well as the configuration restrictions that 
will be imposed by the Maintenance Rule online risk matrix, the impact 
on core damage frequency is negligible.

[[Page 38207]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: P. T. Kuo, Acting.

Westinghouse Electric Corporation (Licensee), Westinghouse Test 
Reactor, Waltz Mill Site, Westmoreland, Pennsylvania, Docket No. 50-22, 
License No. TR-2

    Date of amendment request: December 22, 1997, supplemented on June 
15, 1998.
    Description of amendment request: In 1959, the Westinghouse 
Electric Corporation was granted a license for the Westinghouse Test 
Reactor (WTR) at Waltz Mill. On December 22, 1997, the licensee 
informed the Nuclear Regulatory Commission it had changed its name to 
CBS Corporation, and requested the license to be amended to reflect the 
name change.
    On June 15, 1998, the CBS Corporation agreed that the name of the 
WTR licensee, as reflected on the license, can be revised to ``CBS 
Corporation acting through its Westinghouse Electric Company 
Division.'' Therefore, the purpose of this amendment is to change the 
name of the licensee as indicated on the WTR license from Westinghouse 
Electric Corporation to CBS Corporation acting through its Westinghouse 
Electric Company Division.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). A proposed amendment to an 
operating license for a facility involves no significant hazards 
consideration if operation of the facility in accordance with the 
proposed amendment would not: (1) involve a significant increase in the 
probability or consequences of an accident previously evaluated; or (2) 
create the possibility of a new or different kind of accident from any 
accident previously evaluated; or (3) involve a significant reduction 
in a margin of safety.
    The staff agrees with the licensee's no significant hazards 
consideration determination submitted on June 15, 1998 for the 
following reason.
    This corporate name change does not involve any change in the 
management, organization, location, facilities equipment, or procedures 
related to or personnel responsible for the licensed activities of the 
WTR license. All existing commitments, obligations and representations 
remain in effect.
    Based on a review of the licensee's analysis, and on the staff's 
analysis detailed above, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for license: Lisa A. Campagna, Assistant General Counsel, 
Law Department, CBS Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 
15230.
    NRC Project Director: Seymour H. Weiss.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: February 26, 1998 (TSCR 204).
    Description of amendment request: The proposed changes would modify 
Technical Specifications (TS) and bases to reflect a lower containment 
leakage limit, a revised program for control of primary coolant sources 
outside containment, a revised control room emergency filtration 
design, and the addition of the primary auxiliary building exhaust 
filtration system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for Point Beach include: 
Control rod withdrawal and drop, CVCS [chemical volume control system] 
malfunction (Boron Dilution), startup of an inactive reactor coolant 
loop, reduction in feedwater enthalpy, excessive load increase, losses 
of reactor coolant flow, loss of external electrical load, loss of 
normal feedwater, loss of all AC [alternating current ] power to the 
auxiliaries, turbine overspeed, fuel handling accidents, accidental 
releases of waste liquid or gas, steam generator tube rupture, steam 
pipe rupture, control rod ejection, and primary coolant system 
ruptures.
    This license amendment request proposes to change the limiting 
conditions for operation, action statements, allowable outage times, 
and surveillance requirements for the Point Beach Nuclear Plant [PBNP] 
Technical Specifications associated with the maximum permissible 
containment leak rate, control room emergency filtration, primary 
auxiliary building exhaust filtration, and primary coolant sources 
outside containment. These proposed changes do not cause an increase in 
the probabilities of any accidents previously evaluated because these 
changes will not cause an increase in the probability of any initiating 
events for accidents previously evaluated. In particular, these changes 
affect accident mitigation systems and equipment which do not cause 
accidents.
    The consequences of the accidents previously evaluated in the PBNP 
FSAR [Final Safety Analysis Report] are determined by the results of 
analyses that are based on initial conditions of the plant, the type of 
accident, transient response of the plant, and the operation and 
failure of equipment and systems. The changes proposed in this license 
amendment request provide appropriate limiting conditions for 
operation, action statements, allowable outage times, and surveillance 
requirements for maximum permissible containment leak rate, control 
room emergency filtration, primary auxiliary building exhaust 
filtration, and primary coolant sources outside containment.
    The proposed changes affect components that are required to ensure 
the proper operation of accident mitigation systems and equipment. The 
proposed changes do not increase the probability of failure of this 
equipment or its ability to operate as required for the accidents 
previously evaluated in the PBNP FSAR.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the Point Beach 
Nuclear Plant FSAR, because the factors that are used to determine the 
consequences of accidents are not being changed.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of

[[Page 38208]]

a new or different kind of accident from any accident previously 
evaluated.
    New or different kinds of accidents can only be created by new or 
different accident initiators or sequences. New and different types of 
accidents (different from those that were originally analyzed for Point 
Beach) have been evaluated and incorporated into the licensing basis 
for Point Beach Nuclear Plant. Examples of different accidents that 
have been incorporated into the Point Beach Licensing basis include 
anticipated transients without scram and station blackout. The changes 
proposed by this license amendment request do not create any new or 
different accident initiators or sequences because these changes to 
limiting conditions for operation, action statements, allowable outage 
times, and surveillance requirements for maximum permissible 
containment leak rate, control room emergency filtration, primary 
auxiliary building exhaust filtration, and primary coolant sources 
outside containment will not cause failures of equipment or accident 
sequences different than the accidents previously evaluated. Therefore, 
these proposed Technical Specifications changes do not create the 
possibility of an accident of a different type than any previously 
evaluated in the Point Beach FSAR.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design and 
operation of the reactor and containment and the safety systems that 
provide their protection.
    The changes proposed by this license amendment request provide the 
appropriate limiting conditions for operation, action statements, 
allowable outage times, and surveillance requirements for maximum 
permissible containment leak rate, control room emergency filtration, 
primary auxiliary building exhaust filtration, and primary coolant 
sources outside containment. This ensures that the safety systems that 
protect the reactor and containment will operate as required. The 
design and operation of the reactor and containment are not affected by 
these proposed changes. Therefore, the margins of safety for Point 
Beach are not being reduced because the design and operation of the 
reactor and containment are not being changed and the safety systems 
and limiting conditions of operation for these safety systems that 
provide their protection that are being changed will continue to meet 
the requirements for accident mitigation for Point Beach Nuclear Plant.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: May 28, 1998 (TSCR 203).
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) to provide a specific numerical 
setting for reactor trip, reactor coolant pump trip, and auxiliary 
feedwater initiation on a loss of power to the 4 kilovolt (kV) buses. 
Changes to the bases for the affected TS sections are also being made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant [PBNP] in accordance 
with the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents potentially affected by the proposed amendments 
previously evaluated for Point Beach include losses of reactor coolant 
flow, loss of external electrical load, loss of normal feedwater, and 
loss of all AC [alternating current] power to the auxiliaries.
    This license amendment request proposes to clarify the setting 
limit for the undervoltage reactor trip, auxiliary feedwater initiation 
and reactor coolant pump trip by providing an actual numerical value in 
place of the word ``Normal'' thereby eliminating any confusion as to 
the actual value used in the setting limit for this protection 
function.
    This proposed change does not cause an increase in the 
probabilities of any accidents previously evaluated because the change 
will not cause an increase in the probability of any initiating events 
for accidents previously evaluated. In particular, the proposed change 
more clearly defines the actual setting limit for the 4 KV undervoltage 
protection function taking into account the effects of voltage decay 
and response times. This is a protection function for mitigation of 
these events. Appropriate delay times are implemented in this function 
to ensure momentary voltage transients do not initiate these events 
while ensuring appropriate protection for these loss of power events. 
Therefore, there is no significant increase in the probability or 
consequences of any event previously analyzed.
    The consequences of the accidents previously evaluated in the PBNP 
FSAR [Final Safety Analysis Report] are determined by the results of 
analyses that are based on initial conditions of the plant, the type of 
accident, transient response of the plant, and the operation and 
failure of equipment and systems. The changes proposed in this license 
amendment request provide appropriate limiting conditions for the 
setting limits for the Point Beach Nuclear Plant Technical 
Specifications for the 4 KV undervoltage protection function. Thus the 
analyses of the events remain valid and demonstrate that there are no 
radiological consequences from these events.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the Point Beach 
Nuclear Plant FSAR, because the factors that are used to determine the 
consequences of accidents are not being changed.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different kinds of accidents can only be created by new or 
different accident initiators or sequences. New and different types of 
accidents (different from those that were originally analyzed for Point 
Beach) have been evaluated and incorporated into the licensing basis 
for Point Beach Nuclear Plant. Examples of different accidents that 
have been incorporated into the Point Beach Licensing basis include 
anticipated transients without scram and station blackout.

[[Page 38209]]

    The change proposed by the amendments to provide specific 
undervoltage setting limits does not create any new or different 
accident initiators or sequences because the change to the 4 KV 
undervoltage protection function will not cause failures of equipment 
or accident sequences different than the accidents previously 
evaluated. Therefore, the proposed Technical Specification change does 
not create the possibility of an accident of a different type than any 
previously evaluated in the Point Beach FSAR.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments [will] not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design and 
operation of the reactor and containment and the safety systems that 
provide their protection.
    The change proposed by this license amendment request provides the 
appropriate setting limit for the 4 KV undervoltage protection 
function. This ensures that the safety systems that protect the reactor 
and containment will operate as required. The design and operation of 
the reactor and containment are not affected by these proposed changes. 
Therefore, the margins of safety for Point Beach are not being reduced 
because the design and operation of the reactor and containment are not 
being changed.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: June 17, 1998, as supplemented June 23, 
1998.
    Brief description of amendment request: The proposed amendment 
would revise Section 3.1.1c of the Technical Specifications (TS), 
Appendix A of the Operating License for the Palisades Nuclear Plant, to 
change the minimum required primary coolant system flow. The currently 
specified value is 140.7 x 10 \6\ lb/hr [pounds per hour] or greater, 
when corrected to 532  deg.F. The licensee proposed to revise the TS to 
specify a value of greater than or equal to 352,000 gpm [gallons per 
minute], which is equivalent to approximately 135 x 10 \6\ lb/hr, when 
corrected to 532  deg.F.
    Date of publication of individual notice in Federal Register: July 
2, 1998 (63 FR 36271)
    Expiration date of individual notice: August 3, 1998.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423-3698.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: June 26, 1998 (NRC-98-0040).
    Brief description of amendment request: The proposed amendment 
would provide a one-time extension of the interval for a number of 
technical specification (TS) surveillance requirements that will be 
performed in the sixth refueling outage. TS 4.0.2 and Index page xxii 
would be revised and TS tables 4.0.2-1 and 4.0.2-2 would be replaced to 
reflect the extensions.
    Date of publication of individual notice in Federal Register: July 
2, 1998 (63 FR 36273).
    Expiration date of individual notice: August 3, 1998.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: June 19, 1998.
    Brief description of amendment request: This amendment revises the 
Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and BVPS-2), 
Technical Specifications (TS) definition of a channel calibration to 
add two sentences stating that (1) the calibration of instrument 
channels with resistance temperature detector or thermocouple sensors 
may consist of an inplace qualitative assessment of sensor behavior and 
normal calibration of the remaining adjustable devices in the channel 
and (2) whenever a sensing element is replaced, the next required 
channel calibration shall include an inplace cross calibration that 
compares the other sensing elements with the recently installed sensing 
element. This proposed change would make the BVPS-1 and BVPS-2 TS 
definition of channel calibration consistent with the definition of a 
channel calibration contained in the NRC's improved Standard Technical 
Specifications for Westinghouse Plants (NUREG-1431, Revision 1).
    Date of publication of individual notice in Federal Register: June 
26, 1998.
    Expiration date of individual notice: July 27, 1998 (63 FR 34939).
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna 
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: June 17, 1998.
    Brief description of amendment request: This amendment revises the 
applicability requirement in TS Sections 3.4.2, ``Safety/Relief 
Valves'' (Action c); 4.4.2; 3.3.7.5, ``Accident Monitoring 
Instrumentation'' (TS Table 3.3.7.5-1, Action 80 and 4.3.7.5, 
``Surveillance Requirements,'' Table 4.3.7.5-1 ``Accident Monitoring 
Instrumentation Surveillance Requirements''). The change to the 
referenced TSs adds the following applicability footnote:

    Compliance with these requirements for the ``J'' SRV acoustic 
monitor is not required for the period beginning June 15, 1998, 
until the next unit shutdown of sufficient duration to allow for 
containment entry, not to exceed the 9th refueling and inspection 
outage.


[[Page 38210]]


    Date of publication of individual notice in Federal Register: June 
23, 1998 (63 FR 34200).
    Expiration date of individual notice: July 23, 1998.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: May 27, 1997, as supplemented 
on August 1, 1997, and March 24, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Section 6, ``Administrative Controls,'' to incorporate 
revised organizational titles and delete Unit 1 Facility Operating 
License Condition 2.C.(30)(a). In addition, the amendments change the 
submittal frequency of the Radiological Effluent Release Report from 
semiannually to annually and make several administrative and editorial 
changes.
    Date of issuance: June 26, 1998.
    Effective date: Immediately, to be implemented within 90 days.
    Amendment Nos.: 128 and 113.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Unit 1 Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: July 30 ,1997. The 
August 1, 1997, submittal provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination. The March 24, 1998, submittal changed the scope of the 
initial Federal Register notice. The proposed amendments were renoticed 
on May 20, 1998 (63 FR 27759).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station, 
Unit 1, LaSalle County, Illinois

    Date of application for amendment: November 24, 1997, as 
supplemented April 16, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification 3/4 3.2, ``Isolation Actuation Instrumentation'' to add/
revise various isolation setpoints for leak detection instrumentation. 
These changes are necessary due to modifications to the Reactor Water 
Cleanup (RWCU) system to restore ``hot'' suction to the RWCU pumps and 
due to a re-evaluation of the high energy line break analysis. In 
addition, the amendment eliminates isolation actuation trip functions 
for the Residual Heat Removal system steam condensing mode and shutdown 
cooling mode.
    Date of issuance: July 6, 1998.
    Effective date: Immediately, to be implemented prior to restart 
from L1F35
    Amendment No.: 129.
    Facility Operating License No. NPF-11: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2278). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of application of amendments: May 30, 1997, as supplemented 
May 7, 1998 and June 18, 1998.
    Brief description of amendments: The amendments revise the Facility 
Operating License and Technical Specifications to reflect the 
permanently shut down and defueled status of the reactor.
    Date of issuance: June 30, 1998.
    Effective date: As of the date of issuance (June 30, 1998) and 
shall be implemented within 90 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-61: The amendments revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38132 and 62 FR 38133). The May 7, 1998, supplement relocated the 
provisions of Technical Specification 3/4.9.15. The June 18, 1998, 
supplement consisted of supporting technical information. The 
supplements did not change the staff's initial proposed no significant 
hazards consideration determination or expand the scope of the original 
notice. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 30, 1998.
    No significant hazards consideration received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: January 27, 1998 (Reference NRC-98-
0023).
    Brief description of amendment: This amendment revises the Fermi 1 
License to allow Detroit Edison to receive, acquire, possess, use, and 
transfer byproduct material without restriction to chemical form for 
sample analysis, instrument calibration, or associated

[[Page 38211]]

with radioactive apparatus, hardware, tools, and equipment, provided 
the cumulative radioactive material quantity of the byproduct material 
does not exceed the criteria contained in Section 30.72, Schedule C, 
Quantities of Radioactive Materials Requiring Consideration of the Need 
for an Emergency Plan for Responding to a Release.
    Date of issuance: June 22, 1998.
    Effective date: Within 60 calendar days from the date of issuance 
of this amendment.
    Amendment No.: 12.
    Facility Operating License No. DPR-9: Amendment revised License by 
adding a subpart 3 to Part 2.B.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17223).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

Duke Energy Corporation, Docket Nos. 50-269 and 50-287, Oconee Nuclear 
Station, Units 1 and 3, Seneca, South Carolina

    Date of application of amendments: June 4, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification 4.17.2 to allow continued operation with certain steam 
generator tubes that exceed their repair limit as a result of tube end 
anomalies. This action temporarily exempts these tubes from the 
requirement for sleeving, rerolling, or removal from service until they 
are repaired during or before the next scheduled refueling outages for 
the respective unit. This action supersedes the Notice of Enforcement 
Discretion that was issued by the staff on June 4, 1998.
    Date of Issuance: July 1, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--230; Unit 2--227.
    Facility Operating License Nos. DPR-38 and DPR-55: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. (63 FR 33097 dated June 17, 1998). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by July 16, 1998, but indicated that if the Commission makes a 
final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated July 1, 1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: June 3, 1997, as supplemented by letter 
dated May 1, 1998.
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications (TSs) by changing the action requirements 
for TS 3/4.3.2 for the Safety Injection System Sump Recirculation 
Actuation Signal (RAS). It revised the allowed outage time for a 
channel of RAS to be in the tripped condition from ``prior to entry 
into the applicable MODE(S) following the next COLD SHUTDOWN'' to the 
more restrictive time limit of 48 hours, and added a shutdown 
requirement. Additionally, the TS 3.0.4 exemption was removed from the 
action for the tripped condition. A change to TS Bases Section 3/4.3.2 
was also included.
    Date of issuance: July 2, 1998.
    Effective date: July 2, 1998, to be implemented within 60 days.
    Amendment No.: 143.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications. Date of initial notice in Federal Register: 
June 18, 1997 (62 FR 33124).
    The additional information contained in the supplemental letter 
dated May 1, 1998, was clarifying in nature and thus, it was within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 2, 1998, as supplemented by letter 
dated April 21, 1998.
    Description of amendment request: The amendment revised Technical 
Specification 4.5.2.b.1 for the emergency core cooling system 
subsystems to delete the requirement to vent the operating chemical 
volume and control system centrifugal pump casing.
    Date of issuance: June 24, 1998.
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 58.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17225).
    The supplemental letter provided clarifying information that did 
not change the staff's proposed no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: April 13, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by adding a new TS 3.5.5, ``Emergency Core Cooling 
Systems--Trisodium Phosphate (TSP).'' The TSP surveillance requirements 
in TSs 4.5.2.c.3 and 4.5.2.c.4 are relocated to new TS 3.5.5 as TS 
4.5.5.1 and TS 4.5.5.2, respectively. Also, the amount of TSP is 
increased, the surveillance requirements are modified, a new limiting 
condition of operation is included, and the applicable TS Index pages 
and Bases sections are updated to reflect the changes.
    Date of issuance: June 22, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 217.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25114).

[[Page 38212]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: November 13, 1997, as 
supplemented on December 29, 1997, and April 8, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by modifying TS 3.1.2.1, ``Flow Paths--Shutdown;'' 
TS 3.1.2.2, ``Flow Paths--Operating;'' TS 3.1.2.3, ``Charging Pump--
Shutdown;'' TS 3.1.2.4, ``Charging Pumps--Operating;'' TS 3.1.2.5, 
``Boric Acid Pumps--Shutdown;'' TS 3.1.2.6, ``Boric Acid Pumps--
Operating;'' TS 3.1.2.8, ``Borated Water Sources--Operating;'' TS 
3.4.1.3, ``Coolant Loops and Coolant Circulation--Shutdown;'' TS 3.4.3, 
``Relief Valves;'' TS 3.4.9.1, ``Reactor Coolant System;'' TS 3.4.9.2, 
``Pressurizer;'' TS 3.4.9.3, ``Overpressure Protection Systems;'' TS 
3.5.3, ``ECCS Subsystems--T<INF>avg</INF> < 300  deg.F;'' and TS 
3.10.3, ``Pressure/Temperature Limitation--Reactor Criticality,'' and 
their associated Bases in the areas that are affected by the modified 
Low Temperature Overpressure Protection system, the updated reactor 
coolant system pressure and temperature curves and heatup and cooldown 
limits. Additionally, minor changes are made to correct various items, 
such as, updating of redundant or outdated TSs.
    Date of issuance: July 1, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 218.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4315).
    The December 29, 1997, and April 8, 1998, letters provided 
clarifying information that did not change the scope of the November 
13, 1997, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: February 27, 1997, as 
supplemented by letter dated December 4, 1997.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant, (DCPP) Unit Nos. 1 and 2 to change Technical Specification (TS) 
3/4.8.1.1, ``A.C. Sources--Operating,'' to clarify that emergency 
diesel generator (EDG) testing is initiated from standby conditions 
rather than ``ambient'' conditions. The associated TS Bases were 
revised to discuss the temperature range that satisfies EDG standby 
conditions. TS 3/4.3.2, ``Instrumentation--Engineering Safety Features 
Actuation System Instrumentation'' was also changed. This revision 
clarified that when one or both of the first level load shed relays, or 
one or both of the second level undervoltage relays are inoperable, the 
associated EDG for that bus shall be declared inoperable.
    Date of issuance: June 5, 1998.
    Effective date: June 5, 1998, to be implemented within 90 days from 
the date of issuance.
    Amendment Nos.: Unit 1--127; Unit 2--125
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17240).
    The December 4, 1997, supplemental letter provided additional 
clarifying information and did not change the staff's initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated June 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: April 10, 1998, as supplemented 
by letter dated May 1, 1998.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Units 1 and 2 to revise TS 6.2.2.g and TS 6.3 to change 
the name of the Operations Manager to Operations Director, to add the 
position of Operations Middle Manager, and to change the requirement 
for the Operations Director to hold a senior reactor operator (SRO) 
license.
    Date of issuance: June 11, 1998.
    Effective date: June 11, 1998.
    Amendment Nos.: Unit 1-128; Unit 2-126.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25116).
    The May 1, 1998, supplemental letter provided additional clarifying 
information and did not change the staff's initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
June 11, 1998.
    No significant hazards consideration comments received: Yes.
    The Commission received one letter with comments which did not 
change its finding and conclusion as discussed in the safety 
evaluation.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: May 12, 1998
    Brief description of amendments: These amendments relocate certain 
requirements related to fire protection from the TSs to the Updated 
Final Safety Analysis Report. The TS sections to be relocated are: 3/
4.3.7.9, Fire Detection Instrumentation; 3/4.7.6, Fire Suppression 
Systems; 3/4.7.7, Fire

[[Page 38213]]

Rated Assemblies; and 6.2.2e, Fire Brigade Staffing. The amendments 
also replace License Condition 2.C.(6) for Unit 1 and License Condition 
2.C.(3) for Unit 2. These amendments are consistent with the guidance 
of NRC Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
from Technical Specifications.''
    Date of issuance: June 24, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 177 and 150.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications and Licenses.
    Date of initial notice in Federal Register: May 21, 1998 (63 FR 
28010).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: November 26, 1997, as 
supplemented April 17, 1998.
    Brief description of amendment: The amendment relocates snubber 
operability, surveillance, and records requirements from the Technical 
Specifications to plant controlled documents. A condition is added to 
the license to require that the relocated requirements be described in 
the Final Safety Analysis Report such that 10 CFR 50.59 will apply to 
future changes to those requirements.
    Date of issuance: June 30, 1998.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 243.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4352).
    The April 17, 1998, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of application for amendments: November 6, 1995, as 
supplemented by letters dated January 9, 1998, and February 3, 1998, 
for the safety injection tanks (SITs), and November 8, 1995, as 
supplemented by letters dated January 9, 1998, and February 3, 1998, 
for the low pressure safety injection (LPSI).
    Brief description of amendments: The amendments modify the 
technical specifications (TSs) to extend the allowed outage times 
(AOTs) for a single inoperable SIT from one hour to 24 hours, and for a 
single inoperable SIT specifically due to malfunctioning SIT water 
level or nitrogen cover pressure instrumentation inoperability from one 
hour to 72 hours. In addition, the amendments extend the AOT for a 
single inoperable LPSI train from 72 hours to 7 days. The amendments 
also add a Configuration Risk Management Program to the TSs that puts a 
proceduralized probabilistic risk assessment-informed process in place 
that ensures the licensee assesses the overall impact of plant 
maintenance on plant risk.
    Date of issuance: June 19, 1998.
    Effective date: June 19, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2--139; Unit 3--131.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15995) and February 11, 1998 (63 FR 6991).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 19, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Southern Nuclear Power Company, Inc., et al., Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: September 4, 1997, as 
supplemented by letters dated November 20, 1997, May 19 and June 12, 
1998.
    Brief description of amendments: The changes to the common 
Technical Specifications allow an increase in the Unit 1 spent fuel 
storage capacity from 288 to 1476 fuel assemblies.
    Date of issuance: June 29, 1998.
    Effective date: As of the date of issuance to be implemented on a 
schedule consistent with the receipt and storage of new fuel in the 
fall of 1998 for the spring 1999 refueling outage of Unit 1.
    Amendment Nos.: Unit 1--102; Unit 2--80.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications, Operating Licenses, and Appendix 
D.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68317); and renoticed on May 11, 1998 (63 FR 25883).
    The supplements dated May 19 and June 12, 1998, provided clarifying 
information that did not change the scope of the September 4, 1997, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 7, 1996, as supplemented 
on September 26, 1997, January 21, 1998, May 28, 1998, and June 29, 
1998 (TS 95-19).
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by relocating portions of Section 6, 
``Administrative Controls,'' to the Sequoyah Nuclear Quality Assurance 
Plan. This Change is consistent with NUREG-1431, ``Standard Technical 
Specifications--Westinghouse Plants.''
    Date of issuance: July 1, 1998.
    Effective date: July 1, 1998.
    Amendment Nos.: 233 and 223.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37302).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 1, 1998.
    No significant hazards consideration comments received: None.

[[Page 38214]]

    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: December 23, 1997, as 
supplemented by letter dated June 11, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Section 1.0, ``Definitions,'' to clarify the meaning 
of core alteration; relocates TS Section 3/4.9.5, ``Refueling 
Operations--Communications,'' and the associated bases to the Technical 
Requirements Manual; and adds TS Section 3.0.6 and associated bases to 
address the return to service of inoperable equipment.
    Date of issuance: June 30, 1998.
    Effective date: June 30, 1998.
    Amendment No.: 224.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4327).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 1998. .
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: March 25, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Sections 6.1.1; 6.2.1.b; 6.5.1.1; 
6.5.1.6.a,d,h, and m; 6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b; 6.2.3.5; 
6.5.1.2; 6.5.1.7.a for Unit 1 and 6.1.1; 6.2.1.b; 6.5.1.1; 
6.5.1.6.a,d,h, and m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b; 6.2.3.5; 
6.5.1.2; and 6.5.1.7.a for Unit 2, changing the title of Station 
Manager to Site Vice President, and the titles of the Assistant Station 
Managers to Manager-Station Operation and Maintenance and Manager-
Station Safety and Licensing.
    Date of issuance: June 23, 1998.
    Effective date: June 23, 1998.
    Amendment Nos.: 212 and 193.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19980).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

    Dated at Rockville, Maryland, this 8th day of July 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-18684 Filed 7-14-98; 8:45 am]
BILLING CODE 7590-01-P